Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving No Significant Hazards Considerations, 28891-28905 [2016-10949]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 81, Number 90 (Tuesday, May 10, 2016)]
[Notices]
[Pages 28891-28905]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-10949]


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NUCLEAR REGULATORY COMMISSION

[NRC-2016-0093]


Applications and Amendments to Facility Operating Licenses and 
Combined Licenses Involving No Significant Hazards Considerations

AGENCY: Nuclear Regulatory Commission.

ACTION: Biweekly notice.

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SUMMARY: Pursuant to Section 189a. (2) of the Atomic Energy Act of 
1954, as amended (the Act), the U.S. Nuclear Regulatory Commission 
(NRC) is publishing this regular biweekly notice. The Act requires the 
Commission to publish notice of any amendments issued, or proposed to 
be issued, and grants the Commission the authority to issue and make 
immediately effective any amendment to an operating license or combined 
license, as applicable, upon a determination by the Commission that 
such amendment involves no significant hazards consideration, 
notwithstanding the pendency before the Commission of a request for a 
hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 12 to April 25, 2016. The last 
biweekly notice was published on April 26, 2016 (81 FR 24659).

DATES: Comments must be filed by June 9, 2016. A request for a hearing 
must be filed by July 11, 2016.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific subject):
     Federal Rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2016-0093. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Mail comments to: Cindy Bladey, Office of Administration, 
Mail Stop: OWFN-12-H08, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001.
    For additional direction on obtaining information and submitting 
comments, see ``Obtaining Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: Shirley Rohrer, Office of Nuclear

[[Page 28892]]

Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-5411, email: Shirley.Rohrer@nrc.gov.

I. Obtaining Information and Submitting Comments

A. Obtaining Information

    Please refer to Docket ID NRC-2016-0093 when contacting the NRC 
about the availability of information for this action. You may obtain 
publicly-available information related to this action by any of the 
following methods:
     Federal rulemaking Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2016-0093.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to pdr.resource@nrc.gov. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in the 
SUPPLEMENTARY INFORMATION section.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

B. Submitting Comments

    Please include Docket ID NRC-2016-0093, facility name, unit 
number(s), application date, and subject in your comment submission.
    The NRC cautions you not to include identifying or contact 
information that you do not want to be publicly disclosed in your 
comment submission. The NRC will post all comment submissions at https://www.regulations.gov as well as enter the comment submissions into 
ADAMS. The NRC does not routinely edit comment submissions to remove 
identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment into ADAMS.

II. Notice of Consideration of Issuance of Amendments to Facility 
Operating Licenses and Combined Licenses and Proposed No Significant 
Hazards Consideration Determination

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in Sec.  50.92 of title 10 of the Code of 
Federal Regulations (10 CFR), this means that operation of the facility 
in accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, (2) create the possibility of a new or different 
kind of accident from any accident previously evaluated, or (3) involve 
a significant reduction in a margin of safety. The basis for this 
proposed determination for each amendment request is shown below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.

A. Opportunity to Request a Hearing and Petition for Leave to Intervene

    Within 60 days after the date of publication of this notice, any 
person(s) whose interest may be affected by this action may file a 
request for a hearing and a petition to intervene with respect to 
issuance of the amendment to the subject facility operating license or 
combined license. Requests for a hearing and a petition for leave to 
intervene shall be filed in accordance with the Commission's ``Agency 
Rules of Practice and Procedure'' in 10 CFR part 2. Interested 
person(s) should consult a current copy of 10 CFR 2.309, which is 
available at the NRC's PDR, located at One White Flint North, Room O1-
F21, 11555 Rockville Pike (first floor), Rockville, Maryland 20852. The 
NRC's regulations are accessible electronically from the NRC Library on 
the NRC's Web site at https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition for leave to intervene is 
filed within 60 days, the Commission or a presiding officer designated 
by the Commission or by the Chief Administrative Judge of the Atomic 
Safety and Licensing Board Panel, will rule on the request and/or 
petition; and the Secretary or the Chief Administrative Judge of the 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address, and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestor's/petitioner's interest. The petition must 
also set forth the specific contentions which the requestor/petitioner 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
requestor/petitioner shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the requestor/
petitioner intends to rely in proving the contention at the hearing. 
The requestor/petitioner must also provide references to those

[[Page 28893]]

specific sources and documents of which the petitioner is aware and on 
which the requestor/petitioner intends to rely to establish those facts 
or expert opinion. The petition must include sufficient information to 
show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the requestor/petitioner to relief. 
A requestor/petitioner who fails to satisfy these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing with respect to resolution of that person's admitted 
contentions, including the opportunity to present evidence and to 
submit a cross-examination plan for cross-examination of witnesses, 
consistent with NRC regulations, policies and procedures.
    Petitions for leave to intervene must be filed no later than 60 
days from the date of publication of this notice. Requests for hearing, 
petitions for leave to intervene, and motions for leave to file new or 
amended contentions that are filed after the 60-day deadline will not 
be entertained absent a determination by the presiding officer that the 
filing demonstrates good cause by satisfying the three factors in 10 
CFR 2.309(c)(1)(i)-(iii). If a hearing is requested, and the Commission 
has not made a final determination on the issue of no significant 
hazards consideration, the Commission will make a final determination 
on the issue of no significant hazards consideration. The final 
determination will serve to decide when the hearing is held. If the 
final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, then any hearing 
held would take place before the issuance of any amendment unless the 
Commission finds an imminent danger to the health or safety of the 
public, in which case it will issue an appropriate order or rule under 
10 CFR part 2.
    A State, local governmental body, federally-recognized Indian 
Tribe, or agency thereof, may submit a petition to the Commission to 
participate as a party under 10 CFR 2.309(h)(1). The petition should 
state the nature and extent of the petitioner's interest in the 
proceeding. The petition should be submitted to the Commission by July 
11, 2016. The petition must be filed in accordance with the filing 
instructions in the ``Electronic Submissions (E-Filing)'' section of 
this document, and should meet the requirements for petitions for leave 
to intervene set forth in this section, except that under Sec.  
2.309(h)(2) a State, local governmental body, or Federally-recognized 
Indian Tribe, or agency thereof does not need to address the standing 
requirements in 10 CFR 2.309(d) if the facility is located within its 
boundaries. A State, local governmental body, Federally-recognized 
Indian Tribe, or agency thereof may also have the opportunity to 
participate under 10 CFR 2.315(c).
    If a hearing is granted, any person who does not wish, or is not 
qualified, to become a party to the proceeding may, in the discretion 
of the presiding officer, be permitted to make a limited appearance 
pursuant to the provisions of 10 CFR 2.315(a). A person making a 
limited appearance may make an oral or written statement of position on 
the issues, but may not otherwise participate in the proceeding. A 
limited appearance may be made at any session of the hearing or at any 
prehearing conference, subject to the limits and conditions as may be 
imposed by the presiding officer. Persons desiring to make a limited 
appearance are requested to inform the Secretary of the Commission by 
July 11, 2016.

B. Electronic Submissions (E-Filing)

    All documents filed in NRC adjudicatory proceedings, including a 
request for hearing, a petition for leave to intervene, any motion or 
other document filed in the proceeding prior to the submission of a 
request for hearing or petition to intervene, and documents filed by 
interested governmental entities participating under 10 CFR 2.315(c), 
must be filed in accordance with the NRC's E-Filing rule (72 FR 49139; 
August 28, 2007). The E-Filing process requires participants to submit 
and serve all adjudicatory documents over the internet, or in some 
cases to mail copies on electronic storage media. Participants may not 
submit paper copies of their filings unless they seek an exemption in 
accordance with the procedures described below.
    To comply with the procedural requirements of E-Filing, at least 
ten 10 days prior to the filing deadline, the participant should 
contact the Office of the Secretary by email at hearing.docket@nrc.gov, 
or by telephone at 301-415-1677, to request (1) a digital 
identification (ID) certificate, which allows the participant (or its 
counsel or representative) to digitally sign documents and access the 
E-Submittal server for any proceeding in which it is participating; and 
(2) advise the Secretary that the participant will be submitting a 
request or petition for hearing (even in instances in which the 
participant, or its counsel or representative, already holds an NRC-
issued digital ID certificate). Based upon this information, the 
Secretary will establish an electronic docket for the hearing in this 
proceeding if the Secretary has not already established an electronic 
docket.
    Information about applying for a digital ID certificate is 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals/getting-started.html. System requirements for accessing 
the E-Submittal server are detailed in the NRC's ``Guidance for 
Electronic Submission,'' which is available on the agency's public Web 
site at https://www.nrc.gov/site-help/e-submittals.html. Participants 
may attempt to use other software not listed on the Web site, but 
should note that the NRC's E-Filing system does not support unlisted 
software, and the NRC Meta System Help Desk will not be able to offer 
assistance in using unlisted software.
    If a participant is electronically submitting a document to the NRC 
in accordance with the E-Filing rule, the participant must file the 
document using the NRC's online, Web-based submission form. In order to 
serve documents through the Electronic Information Exchange System, 
users will be required to install a Web browser plug-in from the NRC's 
Web site. Further information on the Web-based submission form, 
including the installation of the Web browser plug-in, is available on 
the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html.
    Once a participant has obtained a digital ID certificate and a 
docket has been created, the participant can then submit a request for 
hearing or petition for leave to intervene. Submissions should be in 
Portable Document Format (PDF) in accordance with NRC guidance 
available on the NRC's public Web site at https://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the time the 
documents are submitted through the NRC's E-Filing system. To be 
timely, an electronic

[[Page 28894]]

filing must be submitted to the E-Filing system no later than 11:59 
p.m. Eastern Time on the due date. Upon receipt of a transmission, the 
E-Filing system time-stamps the document and sends the submitter an 
email notice confirming receipt of the document. The E-Filing system 
also distributes an email notice that provides access to the document 
to the NRC's Office of the General Counsel and any others who have 
advised the Office of the Secretary that they wish to participate in 
the proceeding, so that the filer need not serve the documents on those 
participants separately. Therefore, applicants and other participants 
(or their counsel or representative) must apply for and receive a 
digital ID certificate before a hearing request/petition to intervene 
is filed so that they can obtain access to the document via the E-
Filing system.
    A person filing electronically using the NRC's adjudicatory E-
Filing system may seek assistance by contacting the NRC Meta System 
Help Desk through the ``Contact Us'' link located on the NRC's public 
Web site at https://www.nrc.gov/site-help/e-submittals.html, by email to 
MSHD.Resource@nrc.gov, or by a toll-free call at 1-866-672-7640. The 
NRC Meta System Help Desk is available between 8 a.m. and 8 p.m., 
Eastern Time, Monday through Friday, excluding government holidays.
    Participants who believe that they have a good cause for not 
submitting documents electronically must file an exemption request, in 
accordance with 10 CFR 2.302(g), with their initial paper filing 
requesting authorization to continue to submit documents in paper 
format. Such filings must be submitted by: (1) First class mail 
addressed to the Office of the Secretary of the Commission, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-0001, Attention: 
Rulemaking and Adjudications Staff; or (2) courier, express mail, or 
expedited delivery service to the Office of the Secretary, Sixteenth 
Floor, One White Flint North, 11555 Rockville Pike, Rockville, 
Maryland, 20852, Attention: Rulemaking and Adjudications Staff. 
Participants filing a document in this manner are responsible for 
serving the document on all other participants. Filing is considered 
complete by first-class mail as of the time of deposit in the mail, or 
by courier, express mail, or expedited delivery service upon depositing 
the document with the provider of the service. A presiding officer, 
having granted an exemption request from using E-Filing, may require a 
participant or party to use E-Filing if the presiding officer 
subsequently determines that the reason for granting the exemption from 
use of E-Filing no longer exists.
    Documents submitted in adjudicatory proceedings will appear in the 
NRC's electronic hearing docket which is available to the public at 
https://ehd1.nrc.gov/ehd/, unless excluded pursuant to an order of the 
Commission, or the presiding officer. Participants are requested not to 
include personal privacy information, such as social security numbers, 
home addresses, or home phone numbers in their filings, unless an NRC 
regulation or other law requires submission of such information. 
However, in some instances, a request to intervene will require 
including information on local residence in order to demonstrate a 
proximity assertion of interest in the proceeding. With respect to 
copyrighted works, except for limited excerpts that serve the purpose 
of the adjudicatory filings and would constitute a Fair Use 
application, participants are requested not to include copyrighted 
materials in their submission.
    For further details with respect to these license amendment 
applications, see the application for amendment which is available for 
public inspection in ADAMS and at the NRC's PDR. For additional 
direction on accessing information related to this document, see the 
``Obtaining Information and Submitting Comments'' section of this 
document.

Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power 
Station, Unit No. 1 (MPS1), New London County, Connecticut

    Date of amendment request: March 28, 2014. A publicly-available 
version is in the ADAMS under Accession No. ML14093A028.
    Description of amendment request: The amendment would make changes 
to the MPS1 Permanently Defueled Technical Specifications (PDTSs) by 
deleting the Table of Contents section and making administrative 
changes to the PDTSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1) Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes are administrative in nature. The proposed 
changes remove the PDTS Table of Contents section and make two other 
administrative changes to the PDTSs. Furthermore, MPS1 has 
permanently ceased operation and is being maintained in a defueled 
condition. Therefore, the only credible design basis accident is a 
fuel handling accident. The administrative changes proposed herein 
are not initiators of any fuel handling accident previously 
evaluated, and, consequently, the probability and consequences of a 
fuel handling accident previously evaluated is not significantly 
increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2) Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes are administrative in nature, therefore no 
new or different accidents result from the proposed changes. The 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed), a change in 
the method of plant operation, or new operator actions. The changes 
do not alter assumptions made in the safety analysis.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3) Do the proposed changes involve a significant reduction in 
the margin of safety?
    Response: No.
    The proposed administrative changes do not involve a change in 
the method of plant operation, do not affect any accident analyses, 
and do not relax any safety system settings.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lillian M. Cuoco, Senior Counsel, Dominion 
Resource Services, Inc., 120 Tredegar Street, RS-2, Richmond, VA 23219.
    NRC Branch Chief: Bruce A. Watson.

Duke Energy Carolinas, LLC, Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of amendment request: February 18, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16076A413.
    Description of amendment request: The amendment would allow a one-
time extension to the 10-year frequency of the McGuire Nuclear Station, 
Units 1

[[Page 28895]]

and 2, containment leakage rate tests. The change would extend the 
period from 10 years to 10.5 years for each unit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to the Technical Specifications (TS) 
involves the extension of the McGuire Nuclear Station (MNS) Type A 
containment integrated leak rate test interval to 10.5 years. The 
current Type A test interval of 120 months (10 years) would be 
extended on a one-time basis to no longer than 10.5 years from the 
last Type A test. This extension is bounded by the 15 month 
extension, permissible only for non-routine emergent conditions, 
allowed in accordance with NEI [Nuclear Energy Institute] 94-01 
revision 0. The proposed extension also does not change the test 
method or procedure. The containment is designed to provide an 
essentially leak tight barrier against the uncontrolled release of 
radioactivity to the environment for postulated accidents. The 
containment and the testing requirements invoked to periodically 
demonstrate the integrity of the containment exist to ensure the 
plant's ability to mitigate the consequences of an accident, and do 
not involve the prevention or identification of any precursors of an 
accident. The change in dose risk for changing the Type A test 
frequency from 10 years to 10.5 years, measured, as an increase to 
the total integrated plant risk for those accident sequences 
influenced by Type A testing, is 0.023 person-rem/year. EPRI 
[Electric Power Research Institute] Report No. 1009325, Revision 2-A 
states that a very small population dose is defined as an increase 
of <= 1.0 person-rem per year, or <= 1% of the total population 
dose, whichever is less restrictive for the risk impact assessment 
of the extended ILRT [integrated leak rate test] intervals. 
Therefore, this proposed extension does not involve a significant 
increase in the probability of an accident previously evaluated.
    As documented in NUREG-1493, Performance-Based Containment Leak-
Test Program, Type B and C tests have identified a very large 
percentage of containment leakage paths, and the percentage of 
containment leakage paths that are detected only by Type A testing 
is very small. The MNS Type A test history supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and; (2) time based as previously discussed. Activity based failure 
mechanisms are defined as degradation due to system and/or component 
modifications or maintenance. Local leak rate test requirements and 
administrative controls such as configuration management and 
procedural requirements for system restoration ensure that 
containment integrity is not degraded by plant modifications or 
maintenance activities. The design and construction requirements of 
the containment combined with the containment inspections performed 
in accordance with ASME Section XI, the Maintenance Rule, and TS 
requirements serve to provide a high degree of assurance that the 
containment would not degrade in a manner that is detectable only by 
a Type A test. Based on the above, the proposed extensions do not 
significantly increase the consequences of an accident previously 
evaluated.
    Therefore, the proposed change does not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed amendment to the TS involves the extension of the 
MNS Type A containment integrated leak rate test interval from 10 
years to 10.5 years. The current Type A test interval of 120 months 
(10 years) would be extended on a one-time basis to 10.5 years from 
the last Type A test. The containment and the testing requirements 
to periodically demonstrate the integrity of the containment exist 
to ensure the plant's ability to mitigate the consequences of an 
accident do not involve any accident precursors or initiators. The 
proposed change does not involve a physical change to the plant 
(i.e., no new or different type of equipment will be installed) or a 
change to the manner in which the plant is operated or controlled.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response: No.
    The proposed amendment to TS 5.5.2 involves the extension of the 
MNS Type A containment integrated leak rate test interval to 10.5 
years. The current Type A test interval of 120 months (10 years) 
would be extended on a one-time basis to no longer than 10.5 years 
from the last Type A test. This amendment does not alter the manner 
in which safety limits, limiting safety system set points, or 
limiting conditions for operation are determined. The specific 
requirements and conditions of the TS Containment Leak Rate Testing 
Program exist to ensure that the degree of containment structural 
integrity and leak tightness that is considered in the plant safety 
analysis is maintained. The overall containment leak rate limit 
specified by TS is maintained.
    The proposed change involves only the extension of the interval 
between Type A containment leak rate tests for MNS. The proposed 
surveillance interval extension is bounded by the 15-year ILRT 
interval currently authorized within NEI 94-01, Revisions 2-A and 3-
A. Industry experience supports the conclusion that Type B and C 
testing detects a large percentage of containment leakage paths and 
that the percentage of containment leakage paths that are detected 
only by Type A testing is small. The containment inspections 
performed in accordance with ASME Section XI, and TS serve to 
provide a high degree of assurance that the containment would not 
degrade in a manner that is detectable only by Type A testing. The 
combination of these factors ensures that the margin of safety in 
the plant safety analysis is maintained. The design, operation, 
testing methods and acceptance criteria for Type A, B, and C 
containment leakage tests specified in applicable codes and 
standards would continue to be met, with the approval of this 
proposed change, since these are not affected by changes to the Type 
A test intervals.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Lara S. Nichols, Deputy General Counsel, 
Duke Energy Corporation, 526 South Church Street--EC07H, Charlotte, NC 
28202.
    NRC Branch Chief: Michael T. Markley.

Exelon Generation Company, LLC, Docket No. 50-461, Clinton Power 
Station (CPS), Unit No. 1, DeWitt County, Illinois

    Date of amendment request: January 25, 2016, as supplemented by 
letter dated March 31, 2016. A publicly-available version is in ADAMS 
under Accession Nos. ML16025A182 and ML16076A077.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) to allow a permanent 
extension of the Type ``A'' integrated leak rate testing and Type ``C'' 
leak rate testing frequencies. This request also proposes to delete 
information in TS 5.5.13 regarding a completed requirement to perform 
Type ``C'' testing in 2008.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed activity involves the extension of the Clinton 
Power Station (CPS), Unit 1, Type A containment test interval to

[[Page 28896]]

15 years, and the extension of the Type C test interval to 75 
months. The current Type A test interval of 120 months (10 years) 
would be extended on a permanent basis to no longer than 15 years 
from the last Type A test. The current Type C test interval of 60 
months for selected components would be extended on a performance 
basis to no longer than 75 months. Extensions of up to nine months 
(total maximum interval of 84 months for Type C tests) are 
permissible only for non-routine emergent conditions. The proposed 
extension does not involve either a physical change to the plant or 
a change in the manner in which the plant is operated or controlled. 
The containment is designed to provide an essentially leak tight 
barrier against the uncontrolled release of radioactivity to the 
environment for postulated accidents. As such, the containment and 
the testing requirements invoked to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident, and do not involve the 
prevention or identification of any precursors of an accident.
    The change in dose risk for changing the Type A Integrated Leak 
Rate Test (ILRT) interval from three-per-ten years to once-per-
fifteen-years, measured as an increase to the total integrated dose 
risk for all accident sequences, is 3.80E-03 person-rem/yr using the 
EPRI [Electric Power Research Institute] guidance with the base case 
corrosion included. This change meets both of the related acceptance 
criteria for change in population dose of less than 1.0 person-rem/
yr or less than 1% person-rem/yr. The change in dose risk drops to 
9.37E-04 person-rem/yr when using the EPRI Expert Elicitation 
methodology. The change in dose risk meets both of the related 
acceptance for change in population dose of less than 1.0 person-
rem/yr or less than 1% person-rem/yr. Therefore, this proposed 
extension does not involve a significant increase in the probability 
of an accident previously evaluated.
    In addition, as documented in NUREG-1493, Types B and C tests 
have identified a very large percentage of containment leakage 
paths, and the percentage of containment leakage paths that are 
detected only by Type A testing is very small. The CPS, Unit 1 Type 
A test history supports this conclusion.
    The integrity of the containment is subject to two types of 
failure mechanisms that can be categorized as: (1) Activity based, 
and, (2) time based. Activity based failure mechanisms are defined 
as degradation due to system and/or component modifications or 
maintenance. Local leak rate test requirements and administrative 
controls such as configuration management and procedural 
requirements for system restoration ensure that containment 
integrity is not degraded by plant modifications or maintenance 
activities. The design and construction requirements of the 
containment combined with the containment inspections performed in 
accordance with American Society of Mechanical Engineers (ASME) 
Section XI, and Technical Specifications (TS) requirements serve to 
provide a high degree of assurance that the containment would not 
degrade in a manner that is detectable only by a Type A test. Based 
on the above, the proposed extensions do not significantly increase 
the consequences of an accident previously evaluated.
    The proposed amendment also deletes an exception previously 
granted to allow one-time extension of the ILRT test frequency for 
CPS. This exception was for an activity that has already taken 
place; therefore, this deletion is solely an administrative action 
that does not result in any change in how CPS is operated.
    Therefore, the proposed change does not result in a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed amendment to the TS 5.5.13, ``Primary Containment 
Leakage Rate Testing Program,'' involves the extension of the CPS, 
Unit 1 Type A containment test interval to 15 years and the 
extension of the Type C test interval to 75 months. The containment 
and the testing requirements to periodically demonstrate the 
integrity of the containment exist to ensure the plant's ability to 
mitigate the consequences of an accident.
    The proposed change does not involve a physical change to the 
plant (i.e., no new or different type of equipment will be 
installed) nor does it alter the design, configuration, or change 
the manner in which the plant is operated or controlled beyond the 
standard functional capabilities of the equipment.
    The proposed amendment also deletes an exception previously 
granted to allow one-time extension of the ILRT test frequency for 
CPS. This exception was for an activity that has already taken 
place; therefore, this deletion is solely an administrative action 
that does not result in any change in how CPS is operated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed amendment to TS 5.5.13 involves the extension of 
the CPS, Unit 1 Type A containment test interval to 15 years and the 
extension of the Type C test interval to 75 months for selected 
components. This amendment does not alter the manner in which safety 
limits, limiting safety system set points, or limiting conditions 
for operation are determined. The specific requirements and 
conditions of the TS Containment Leak Rate Testing Program exist to 
ensure that the degree of containment structural integrity and 
leaktightness that is considered in the plant safety analysis is 
maintained. The overall containment leak rate limit specified by TS 
is maintained.
    The proposed change involves the extension of the interval 
between Type A containment leak rate tests and Type C tests for CPS, 
Unit 1. The proposed surveillance interval extension is bounded by 
the 15-year ILRT interval and the 75-month Type C test interval 
currently authorized within NEI [Nuclear Energy Institute] 94-01, 
Revision 3-A. Industry experience supports the conclusion that Type 
B and C testing detects a large percentage of containment leakage 
paths and that the percentage of containment leakage paths that are 
detected only by Type A testing is small. The containment 
inspections performed in accordance with ASME Section Xl, and TS 
serve to provide a high degree of assurance that the containment 
would not degrade in a manner that is detectable only by Type A 
testing. The combination of these factors ensures that the margin of 
safety in the plant safety analysis is maintained. The design, 
operation, testing methods and acceptance criteria for Type A, B, 
and C containment leakage tests specified in applicable codes and 
standards would continue to be met, with the acceptance of this 
proposed change, since these are not affected by changes to the Type 
A and Type C test intervals.
    The proposed amendment also deletes exceptions previously 
granted to allow one time extensions of the ILRT test frequency for 
CPS, Unit 1. This exception was for an activity that has taken 
place; therefore, the deletion is solely an administrative action 
and does not change how CPS is operated and maintained. Thus, there 
is no reduction in any margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Acting Branch Chief: Justin C. Poole.

Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: February 23, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16054A359.
    Description of amendment request: The amendment would revise the 
Technical Specifications to incorporate previously NRC-approved 
Industry/Technical Specification Task Force 439 (TSTF-439), Revision 2, 
``Eliminate Second Completion Times Limiting Time From Discovery of 
Failure To Meet an LCO.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 28897]]

consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change eliminates certain Completion Times from the 
Technical Specifications. Completion Times are not an initiator to 
any accident previously evaluated. As a result, the probability of 
an accident previously evaluated is not affected. The consequences 
of an accident during the revised Completion Time are no different 
than the consequences of the same accident during the existing 
Completion Times. As a result, the consequences of an accident 
previously evaluated are not affected by this change. The proposed 
change does not alter or prevent the ability of SSCs [systems, 
structures, and components] from performing their intended function 
to mitigate the consequences of an initiating event within the 
assumed acceptance limits. The proposed change does not affect the 
source term, containment isolation, or radiological release 
assumptions used in evaluating the radiological consequences of an 
accident previously evaluated. Further, the proposed change does not 
increase the types or amounts of radioactive effluent that may be 
released offsite, nor significantly increase individual or 
cumulative occupational/public radiation exposures. The proposed 
change is consistent with the safety analysis assumptions and 
resultant consequences. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed change does not involve a physical alteration of 
the plant (i.e., no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. The proposed change does not alter any assumptions made 
in the safety analysis. Therefore, the proposed change does not 
create the possibility of anew or different kind of accident from 
any accident previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed change to delete the second Completion Time does 
not alter the manner in which safety limits, limiting safety system 
settings or limiting conditions for operation are determined. The 
safety analysis acceptance criteria are not affected by this change. 
The proposed change will not result in plant operation in a 
configuration outside of the design basis. Therefore, the proposed 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: Travis L. Tate.

Exelon Generation Company, LLC, Docket Nos. 50-220 and 50-410, Nine 
Mile Point Nuclear Station (NMPNS), Units 1 and 2, Oswego County, New 
York

    Date of amendment request: March 18, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16078A065.
    Description of amendment request: The amendment would revise the 
Technical Specifications (TS) concerning a change to the method of 
calculating core reactivity for the purpose of performing the 
Reactivity Anomalies surveillance at NMPNS, Units 1 and 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed TS changes do not affect any plant systems, 
structures, or components designed for the prevention or mitigation 
of previously evaluated accidents. The amendment would only change 
how the Reactivity Anomalies surveillance is performed. Verifying 
that the core reactivity is consistent with predicted values ensures 
that accident and transient safety analyses remain valid. This 
amendment changes the TS requirements such that, rather than 
performing the surveillance by comparing predicted to actual control 
rod density, the surveillance is performed by a direct comparison of 
keff.
    Therefore, since the Reactivity Anomalies surveillance will 
continue to be performed by a viable method, the proposed amendment 
does not involve a significant increase in the probability or 
consequence of a previously evaluated accident.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    This TS amendment request does not involve any changes to the 
operation, testing, or maintenance of any safety-related, or 
otherwise important to safety systems. All systems important to 
safety will continue to be operated and maintained within their 
design bases. The proposed changes to the Reactivity Anomalies 
surveillance will only provide a new, more efficient method of 
detecting an unexpected change in core reactivity.
    Since all systems continue to be operated within their design 
bases, no new failure modes are introduced and the possibility of a 
new or different kind of accident is not created.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    This proposed TS amendment proposes to change the method for 
performing the Reactivity Anomalies surveillance from a comparison 
of predicted to actual control rod density to a comparison of 
predicted to monitored keff. The direct comparison of 
keff provides a technically superior method of 
calculating any differences in the expected core reactivity. The 
Reactivity Anomalies surveillance will continue to be performed at 
the same frequency as is currently required by the TS, only the 
method of performing the surveillance will be changed. Consequently, 
core reactivity assumptions made in safety analyses will continue to 
be adequately verified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Tamra Domeyer, Associate General Counsel, 
Exelon Generation Company, LLC, 4300 Winfield Road, Warrenville, IL 
60555.
    NRC Branch Chief: Travis L. Tate.

Exelon Generation Company, LLC, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Units 1 and 2, Will County, Illinois and Docket Nos. 
STN 50-454 and STN 50-455, Byron Station, Unit Nos. 1 and 2, Ogle 
County, Illinois

    Date of amendment request: February 23, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16055A149.
    Description of amendment request: The amendment would (1) revise 
Technical Specification (TS) 4.2.1, ``Reactor Core, Fuel Assemblies,'' 
to add Optimized ZIRLO\TM\, as an approved fuel rod cladding material, 
(2) revise TS 5.6.5.b to add the Westinghouse topical reports for 
Optimized ZIRLO\TM\ and ZIRLO[supreg], and (3) revise TS 5.6.5.b with a 
non-technical change to the Reference 11 title (replace a semicolon 
with a period).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:


[[Page 28898]]


    EGC [Exelon Generation Company] has evaluated the proposed 
changes for Braidwood and Byron, using the criteria in 10 CFR 50.92, 
and has determined that the proposed changes do not involve a 
significant hazards consideration. The following information is 
provided to support a finding of no significant hazards 
consideration.
    Criteria
    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would allow the use of Optimized ZIRLO\TM\ 
clad nuclear fuel in the reactors. The NRC approved topical report 
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, ``Optimized ZIRLO\TM\ 
prepared by Westinghouse Electric Company LLC (Westinghouse), 
addresses Optimized ZIRLO\TM\ and demonstrates that Optimized 
ZIRLO\TM\ has essentially the same properties as currently licensed 
ZIRLO[supreg]. The fuel cladding itself is not an accident initiator 
and does not affect accident probability. With the approved 
exemption, use of Optimized ZIRLO\TM\ fuel cladding will continue to 
meet all 10 CFR 50.46 acceptance criteria and, therefore, will not 
increase the consequences of an accident. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    Use of Optimized ZIRLO\TM\ clad fuel will not result in changes 
in the operation or configuration of the facility. Topical Report 
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, demonstrated that the 
material properties of Optimized ZIRLO\TM\ are similar to those of 
standard ZIRLO[supreg]. Therefore, Optimized ZIRLO\TM\ fuel rod 
cladding will perform similarly to those fabricated from standard 
ZIRLO[supreg] thus precluding the possibility of the fuel cladding 
becoming an accident initiator and causing a new or different type 
of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change will not involve a significant reduction in 
the margin of safety. Topical Report WCAP-12610-P-A & CENPD-404-P-A, 
Addendum 1-A, demonstrated that the material properties of the 
Optimized ZIRLO\TM\ are not significantly different from those of 
standard ZIRLO[supreg]. Optimized ZIRLO\TM\ is expected to perform 
similarly to standard ZIRLO[supreg] for all normal operating and 
accident scenarios, including both loss of coolant accident (LOCA) 
and non-LOCA scenarios. For LOCA scenarios, where the slight 
difference is Optimized ZIRLO\TM\ material properties relative to 
standard ZIRLO[supreg] could have some impact on the overall 
accident scenario, plant-specific LOCA analyses using Optimized 
ZIRLO\TM\ properties will demonstrate that the acceptance criteria 
of 10 CFR 50.46 have been satisfied. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.
    Based on the above, EGC concludes that the proposed amendment to 
allow the use of Optimized ZIRLO\TM\ fuel cladding material does not 
involve a significant hazards consideration under the standards set 
forth in 10 CFR 50.92(c), and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Bradley J. Fewell, Associate General 
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
    NRC Acting Branch Chief: Justin C. Poole.

FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear 
Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: March 15, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16075A411.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.6.2.2, ``Suppression Pool Water 
Level,'' as well as TS surveillance requirements 3.6.2.4.1 and 
3.6.2.4.4 associated with TS 3.6.2.4, ``Suppression Pool Makeup System 
(SPMU),'' to allow installation of the reactor well to steam dryer 
storage pool gate in the upper containment pool (UCP) in MODES 1, 2, 
and 3. The proposed amendment would also create new special operations 
TS 3.10.9, ``Suppression Pool Makeup--MODE 3 Upper Containment Pool 
Drain-Down,'' to allow draining of the reactor well portion of the UCP 
in MODE 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The changes proposed in the license amendment request specify 
different water level requirements in the upper containment pool and 
suppression pool to permit gate installation in MODES 1, 2, and 3, 
and drain-down of the reactor well in MODE 3. The probability of an 
accident previously evaluated is unrelated to the water level in 
these pools, since they are mitigating systems. The operation or 
failure of a mitigating system does not contribute to the occurrence 
of an accident. No active or passive failure mechanisms that could 
lead to an accident are affected by these proposed changes.
    Suppression pool water levels are increased during upper pool 
gate installation in MODES 1, 2, and 3 and during reactor well 
drain-down in MODE 3, with a potential for an increased probability 
of drywell flooding during an inadvertent dump of the upper 
containment pool. An inadvertent dump of the upper pool during any 
period of operation with a pressurized vessel does not represent, in 
and of itself, any significant hazard to the public, the plant 
operating personnel, or any plant equipment. The piping components 
which would be affected in this event have been analyzed for the 
flooding effect, and it has been determined that this event could 
not initiate a loss of coolant accident (LOCA).
    The changes have no impact on the ability of any of the 
emergency core cooling systems (ECCS) to function adequately, since 
adequate net positive suction head (NPSH) is maintained. The 
increase in suppression pool water level to compensate for the 
reduction in UCP volume will provide reasonable assurance that the 
minimum post-accident vent coverage is adequate to assure the 
pressure suppression function of the suppression pool is 
accomplished. The suppression pool water level will be raised above 
the current high water level for the proposed reactor well drain-
down activity only after the reactor pressure has been reduced 
sufficiently to assure that the hydrodynamic loads from a loss of 
coolant accident will not exceed the design values. The reduced 
reactor pressure will also ensure that the loads due to main steam 
safety relief valve actuation with an elevated pool level are within 
the design loads.
    Relative to dose rates on the refuel floor, the resultant dose 
rates from the reactor in MODES 3 and 4 are the same regardless of a 
drain-down of the upper pool reactor well. Relative to a low 
pressure LOCA in MODE 3, the reduced post-LOCA containment pressure 
and the decay time to reach MODE 3 conditions ensures that post-
accident dose consequences are bounded by the design-basis accident 
LOCA.
    Therefore, the proposed amendment does not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from an accident previously evaluated?
    Response: No.
    The proposed changes specify different water level requirements 
in the upper containment pool and suppression pool to permit gate 
installation in MODES 1, 2, and 3, and drain-down of the reactor 
well in MODE 3. These changes do not affect or alter the ability of 
the suppression pool makeup

[[Page 28899]]

(SPMU) system to perform its design function. The proposed change in 
the pool water levels will maintain the design function of 
mitigating the pressure and temperature increase generated by a 
LOCA, and will maintain the required drywell vent coverage during 
post-accident ECCS draw down.
    The altered water levels in the pools do not create a different 
type of accident than presently evaluated. With the reduced pressure 
in the reactor coolant system, the GOTHIC computer program 
simulations demonstrate that the accident responses at defined 
conditions with the reactor well drained in MODE 3 are bounded by 
the current design basis accidents.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to the UCP and the suppression pool water 
levels do not introduce any new setpoints at which protective or 
mitigating actions are initiated. Current instrument setpoints 
remain unaltered by this change. Although the water levels are 
adjusted for the UCP gate installation and the reactor well drain-
down activity, the design and functioning of the containment 
pressure suppression system remains unchanged. The proposed total 
water volume is sufficient to provide high confidence that the 
pressure suppression and containment systems will be capable of 
mitigating large and small break accidents. All analyzed accident 
results remain within the design values for the structures and 
equipment.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: David W. Jenkins, Attorney, FirstEnergy 
Corporation, Mail Stop A-GO-15, 76 South Main Street, Akron, OH 44308.
    NRC Branch Chief: David J. Wrona.

Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo 
County, California

    Date of amendment request: March 23, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16084A588.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.4.12, ``Low Temperature 
Overpressure Protection (LTOP) System,'' to reflect the mass input 
transient analysis that assumes an emergency core cooling system (ECCS) 
centrifugal charging pump (CCP) and the normal charging pump (NCP) 
capable of simultaneously injecting into the reactor coolant system 
(RCS) during TS 3.4.12 applicability.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change revises TS 3.4.12 to allow an ECCS CCP and 
the NCP aligned to LTOP orifice to be capable of injecting into the 
RCS during low RCS pressures and temperatures. The LCO [Limiting 
Condition for Operation] provides RCS overpressure protection by 
having a minimum coolant input capability and have adequate pressure 
relief capability. Analyses have demonstrated that one power 
operated relief valve (PORV) or an RCS vent of at least 2.07 square 
inches is capable of limiting the RCS pressure excursions below the 
10 CFR 50, Appendix G limits for the design basis LTOP limits.
    The proposed change does not adversely affect accident 
initiators or precursors, nor alter the design assumptions, 
conditions, and configuration of the facility or the manner in which 
the plant is operated and maintained. The proposed change does not 
adversely affect the ability of structures, systems, and components 
to perform their intended safety function to mitigate the 
consequences of an initiating event within the assumed acceptance 
limits. The proposed change does not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of any accident previously 
evaluated. Further, the proposed change does not increase the types 
and amounts of radioactive effluent that may be released offsite, 
nor significantly increase individual or cumulative occupational/
public radiation exposure.
    The NRC has previously evaluated a similar LAR [license 
amendment request] related to Wolf Creek Generating Station. In 
Amendment No. 207, the NRC concluded that the proposed change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated [ADAMS Accession 
No. ML13282A534].
    In 2007, PG&E replaced the Unit 1 non-safety-related PDP 
[positive displacement pump] with a non-safety-related CCP, called 
the NCP, in order to alleviate operational issues associated with 
the PDP. In 2008, PG&E performed the replacement on Unit 2. PG&E 
also designed, tested, and installed an FCO [flow choking orifice] 
called the LTOP orifice to be used during LTOP operation to ensure 
that the total maximum mass injection capability with the NCP 
remained bounded by the LTOP mass injection analysis. These changes 
were implemented under 10 CFR 50.59. However, no physical changes 
are being made to the plant as a result of the proposed license 
amendment.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different accident from any accident previously evaluated?
    Response: No.
    The proposed change revises TS 3.4.12 to allow an ECCS CCP and 
the NCP aligned to LTOP orifice to be capable of simultaneously 
injecting into the RCS during low RCS pressures and temperatures. 
The LCO provides RCS overpressure protection by having a minimum 
coolant input capability and have adequate pressure relief 
capability. Analyses have demonstrated that one PORV or an RCS vent 
of at least 2.07 square inches is capable of limiting the RCS 
pressure excursions below the 10 CFR 50, Appendix G limits for the 
design basis LTOP limits.
    The proposed change will not physically alter the plant (no new 
or different type of equipment will be installed) or change the 
methods governing normal plant operation. The proposed change does 
not introduce new accident initiators or impact assumptions made in 
the safety analysis. Testing requirements continue to demonstrate 
that the LCOs are met and the system components are functional.
    Therefore, the proposed change does not create the possibility 
of a new or different accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change does not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. The safety analysis acceptance criteria 
are not impacted by this change. The proposed change will not result 
in plant operation in a configuration outside the design basis.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Attorney for licensee: Jennifer Post, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, CA 94120.
    NRC Branch Chief: Robert J. Pascarelli.

[[Page 28900]]

South Carolina Electric and Gas Company, Docket Nos. 52-027 and 52-028, 
Virgil C. Summer Nuclear Station (VCSNS) Units 2 and 3, Fairfield 
County, South Carolina

    Date of amendment request: March 4, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16067A145.
    Description of amendment request: The proposed changes, if 
approved, would amend Combined License (COL) No. NPF-93 and NPR-94 for 
the VCSNS. The requested amendment proposed changes would depart from 
the approved AP1000 Design Control Document (DCD) ``Tier 2'' and ``Tier 
2*'' information as currently incorporated into the VCSNS Updated Final 
Safety Analysis Report (UFSAR). The changes relate to updating the 
UFSAR text and tables; and information incorporated by reference 
related to Westinghouse Electric Company's Reports WCAP-16096, 
``Software Program Manual for Common QTM Systems,'' (also 
known as the Common Q SPM) Revision 4, WCAP-16097, ``Common Qualified 
Platform Topical Report,'' (also known as the Common Q Topical Report) 
Revision 3, and WCAP-15927, ``Design Process for AP1000 Common Q Safety 
Systems,'' Revision 4; and associated documents and references such as 
a reference to the NRC's Regulatory Guide 1.152, ``Criteria for Use of 
Computers in Safety Systems of Nuclear Power Plants'' (Revision 3, July 
2011), and its associated exceptions. The proposed changes also include 
removal of Tier 2* WCAP-17201-P, ``AC160 High Speed Link Communication 
Compliance to DI&C-ISG-04 Staff Positions 9, 12, 13 and 15 Technical 
Report,'' as a UFSAR incorporated by reference document.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    WCAP-16096 (Common Q Software Program Manual) was updated to 
Revision 4 to reference later NRC endorsed regulatory guides and 
standards and update the requirements for the software design and 
development processes for the Common Q portion of the AP1000 
Protection and Safety Monitoring System (PMS). WCAP-16097 (Common Q 
Topical Report) was updated to Revision 3 to describe new Common Q 
components and standards currently used for the AP1000 PMS 
implementation of the Common Q platform. These two WCAPs have been 
reviewed and approved by the NRC in Safety Evaluations dated 
February 7, 2013. WCAP-15927 was updated to reference the newest 
revisions of WCAP-16096 and WCAP-16097 and for editorial 
corrections. The proposed activity adopts the updated versions as 
incorporated by reference documents into the Updated Final Safety 
Analysis Report. Other proposed document changes support the 
implementation of the updated versions of WCAP-16096, WCAP-16097, 
and WCAP-15927.
    The Common Q platform is an acceptable platform for nuclear 
safety-related applications. The Common Q system meets the 
requirements of 10 CFR part 50, Appendix A, General Design Criteria 
(Criteria 1, 2, 4, 13, 19, 20, 21, 22, 23, 24, and 25), the 
Institute of Electrical and Electronics Engineers (IEEE) Standard 
603-1991 for the design of safety-related reactor protection 
systems, engineered safety features systems and other plant systems, 
and the guidelines of Regulatory Guide 1.152 and supporting industry 
standards for the design of digital systems.
    Because the Common Q platform and the Protection and Safety 
Monitoring System (PMS) implementation of the Common Q platform meet 
the criteria in the applicable General Design Criteria, the 
revisions to these documents do not affect the prevention and 
mitigation of abnormal events, such as accidents, anticipated 
operational occurrences, earthquakes, floods and turbine missiles, 
or their safety or design analyses as described in the licensing 
basis. The incorporation of the updated documents does not adversely 
affect the interface with any structure, system, or component (SSC) 
accident initiator or initiating sequence of events. Thus, the 
probabilities of the accidents previously evaluated in the UFSAR are 
not affected.
    Therefore, the proposed activity does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes to adopt the updated WCAP-16096, WCAP-
16097, and WCAP-15927 into the UFSAR do not adversely affect the 
design or operation of safety-related equipment or equipment whose 
failure could initiate an accident beyond what is already described 
in the licensing basis. These changes do not adversely affect 
fission product barriers. No safety analysis or design basis 
acceptance limit/criterion is challenged or exceeded by the 
requested change.
    Therefore, this activity does not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes to adopt the updated WCAP-16096, WCAP-
16097, and WCAP-15927 into the UFSAR do not adversely affect the 
design, construction, or operation of any plant SSCs, including any 
equipment whose failure could initiate an accident or a failure of a 
fission product barrier. No analysis is adversely affected by the 
proposed changes. Furthermore, no system function, design function, 
or equipment qualification will be adversely affected by the 
changes.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius 
LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514.
    NRC Acting Branch Chief: John McKirgan.

South Carolina Electric and Gas Company, Docket Nos. 52-027 and 52-028, 
Virgil C. Summer Nuclear Station (VCSNS), Units 2 and 3, Fairfield 
County, South Carolina

    Date of amendment request: March 14, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16075A264.
    Description of amendment request: The proposed change would amend 
the Combined License (COL) No. NPF-93 and NPF-94 for the VCSNS. The 
requested amendment proposes to depart from approved AP1000 Design 
Control Document (DCD) Tier 2 information (text, tables, and figures) 
and involved Tier 2* information (as incorporated into the Updated 
Final Safety Analysis Report as plant specific DCD information), and 
also involves a change to the plant-specific Technical Specifications. 
Specifically, the amendment request proposes changes to the plant-
specific AP1000 fuel system design, nuclear design, thermal hydraulic 
design, and accident analyses as described in the licensing basis 
documents. These proposed changes are consistent with those generically 
approved in WCAP-17524-P-A, Revision 1, ``AP1000 Core Reference 
Report.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?

[[Page 28901]]

    Response: No.
    The proposed changes will revise the licensing basis documents 
related to the fuel system design, nuclear design, thermal hydraulic 
design, and accident analyses.
    The UFSAR [Updated Final Safety Analysis Report] Chapter 15 
accident analyses describe the analyses of various design basis 
transients and accidents to demonstrate compliance of the AP1000 
design with the acceptance criteria for these events. The acceptance 
criteria for the various events are based on meeting the relevant 
regulations, general design criteria, the Standard Review Plan, and 
are a function of the anticipated frequency of occurrence of the 
event and potential radiological consequences to the public. As 
such, each design-basis event is categorized accordingly based on 
these considerations. As discussed in Section 5.3 of WCAP-17524-P-A 
Revision 1, the revised accident analyses maintain their plant 
conditions, and thus their frequency designation and consequence 
level as previously evaluated. As confirmed in the Safety Evaluation 
Report (SER), the revised analyses meet the applicable guidelines in 
the Standard Review Plan.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    The proposed changes will revise the licensing basis documents 
related to the fuel system design, nuclear design, thermal hydraulic 
design, and accident analyses.
    The proposed changes would not introduce a new failure mode, 
fault, or sequence of events that could result in a radioactive 
material release. The proposed changes do not alter the design, 
configuration, or method of operation of the plant beyond standard 
functional capabilities of the equipment.
    Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The proposed changes will revise the licensing basis documents 
related to the fuel system design, nuclear design, thermal hydraulic 
design, and accident analyses.
    Safety margins are applied at many levels to the design and 
licensing basis functions and to the controlling values of 
parameters to account for various uncertainties and to avoid 
exceeding regulatory or licensing limits. UFSAR Subsection 4.1.1 
presents the Principle Design Requirements imposed on the fuel and 
control rod mechanism design to ensure that the performance and 
safety criteria described in UFSAR Chapter 4 and Chapter 15 are met. 
The revised fuel system design, nuclear design, thermal hydraulic 
design, and accident analyses maintain the same Principle Design 
Requirements, and further, satisfy the applicable regulations, 
general design criteria, and Standard Review Plan. The effects of 
the changes do not result in a significant reduction in margin for 
any safety function, and were evaluated in the Safety Evaluation 
Report for WCAP-17524-P-A Revision 1 and found to be acceptable.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Kathryn M. Sutton, Morgan, Lewis & Bockius 
LLC, 1111 Pennsylvania Avenue NW, Washington, DC 20004-2514.
    NRC Acting Branch Chief: John McKirgan.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant 
(WBN), Unit 1, Rhea County, Tennessee

    Date of amendment request: February 23, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16054A585.
    Description of amendment request: The amendment would revise the 
WBN Dual Unit Fire Protection Report and would revise the associated 
License Condition regarding the WBN fire protection program. 
Specifically, the amendment requests approval of a deviation from the 
physical separation requirements of 10 CFR part 50, appendix R, section 
III.G.2.d.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    A fire hazards analysis was performed for the areas under the 
scope of this amendment. This fire hazards analysis demonstrates 
that one train of safe shutdown equipment will remain functional in 
the event of an Appendix R fire, even though a radiant energy shield 
will not be provided for two raceway containing safe shutdown 
circuits.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    Response: No.
    A fire hazards analysis was performed for the areas under the 
scope of this amendment. This fire hazards analysis demonstrates 
that one train of safe shutdown equipment will remain functional in 
the event of an Appendix R fire, even though a radiant energy shield 
will not be provided for two raceway containing safe shutdown 
circuits. Based on this, the proposed amendment will not alter the 
requirements or function for systems required during accident 
conditions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    A fire hazards analysis was performed for the areas under the 
scope of this amendment. This fire hazards analysis demonstrates 
that one train of safe shutdown equipment will remain functional in 
the event of an Appendix R fire, even though a radiant energy shield 
will not be provided for two raceway containing safe shutdown 
circuits.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Sherry A. Quirk, Executive Vice President 
and General Counsel, Tennessee Valley Authority, 400 West Summit Hill 
Drive, Knoxville, TN 37902.
    NRC Branch Chief: Benjamin G. Beasley.

III. Previously Published Notices of Consideration of Issuance of 
Amendments to Facility Operating Licenses and Combined Licenses, 
Proposed No Significant Hazards Consideration Determination, and 
Opportunity for a Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

[[Page 28902]]

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of amendment request: March 4, 2016. A publicly-available 
version is in ADAMS under Accession No. ML16064A488.
    Brief description of amendment request: The amendment would revise 
the Cyber Security Plan implementation schedule for Milestone 8 and 
would revise the associated license condition in the Facility Operating 
License.
    Date of publication of individual notice in Federal Register: April 
19, 2016 (81 FR 23011).
    Expiration date of individual notice: May 19, 2016 (public 
comments); June 20, 2016 (hearing requests).

IV. Notice of Issuance of Amendments to Facility Operating Licenses and 
Combined Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR chapter I, which are set 
forth in the license amendment.
    A notice of consideration of issuance of amendment to facility 
operating license or combined license, as applicable, proposed no 
significant hazards consideration determination, and opportunity for a 
hearing in connection with these actions, was published in the Federal 
Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.22(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items can be accessed as described in the 
``Obtaining Information and Submitting Comments'' section of this 
document.

DTE Electric Company, Docket No. 50-341, Fermi 2, Monroe County, 
Michigan

    Date of amendment request: September 24, 2015.
    Brief description of amendment: The amendment revises Surveillance 
Requirements (SRs) to verify that the system locations susceptible to 
gas accumulation are sufficiently filled with water and to provide 
allowances which permit performance of the verification. The changes 
address the concerns discussed in NRC Generic Letter (GL) 2008-01, 
``Managing Gas Accumulation in Emergency Core Cooling, Decay Heat 
Removal, and Containment Spray Systems,'' as described in NRC-approved 
Technical Specifications Task Force (TSTF)-523, Revision 2, ``Generic 
Letter 2008-01, Managing Gas Accumulation.''
    Date of issuance: April 20, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 204. A publicly-available version is in ADAMS under 
Accession. No. ML16069A006; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Facility Operating License No. NPF-43: This amendment revises the 
Facility Operating License and Technical Specifications.
    Date of initial notice in Federal Register: January 5, 2016 (81 FR 
260).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 20, 2016.
    No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of amendment request: April 30, 2015, as supplemented by 
letter dated February 19, 2016.
    Brief description of amendments: The amendments approved adoption 
of an emergency action level scheme based on Nuclear Energy Institute 
(NEI) 99-01, Revision 6, ``Development of Emergency Action Levels for 
Non-Passive Reactors,'' for the Catawba Nuclear Station, Units 1 and 2.
    Date of issuance: April 18, 2016.
    Effective date: As of the date of issuance and shall be implemented 
by March 10, 2017.
    Amendment Nos.: 279 for Unit 1 and 275 for Unit 2. A publicly-
available version is in ADAMS under Accession No. ML16082A038; 
documents related to these amendments are listed in the Safety 
Evaluation enclosed with the amendments.
    Renewed Facility Operating License Nos. NPF-35 and NPF-52: The 
amendments revised the Renewed Facility Operating License.
    Date of initial notice in Federal Register: June 23, 2015 (80 FR 
35980). The supplemental letter dated February 19, 2016, provided 
additional information that clarified the application, did not expand 
the scope of the application as originally noticed, and did not change 
the staff's original proposed no significant hazards consideration 
determination as published in the Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 18, 2016.
    No significant hazards consideration comments received: No.

Duke Energy Carolinas, LLC, Docket Nos. 50-369, 50-370, 50-413, and 50-
414, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North 
Carolina and Catawba Nuclear Station, Units 1 and 2, York County, SC

    Date of amendment request: June 23, 2015.
    Brief description of amendments: The amendments remove superseded 
TS requirements.
    Date of issuance: April 8, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment Nos.: 283, 262, 278, and 274. A publicly-available 
version is in ADAMS under Accession No. ML16060A229; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. NPF-9, NPF-17, NPF-35, and NPF-52: 
Amendments revised the Facility Operating Licenses and Technical 
Specifications.
    Date of initial notice in Federal Register: August 4, 2015 (80 FR 
46347).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 8, 2016.
    No significant hazards consideration comments received: No.

Duke Energy Progress, Inc., Docket No. 50-400, Shearon Harris Nuclear 
Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

    Date of amendment request: April 30, 2015, as supplemented by 
letters dated November 19, 2015, and January 28, 2016.
    Brief description of amendment: The amendment adopted the NRC-
endorsed

[[Page 28903]]

Nuclear Energy Institute (NEI) 99-01, Revision 6, ``Methodology for the 
Development of Emergency Action Levels for Non-Passive Reactors.''
    Date of issuance: April 13, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days of issuance.
    Amendment No.: 149. A publicly-available version is in ADAMS under 
Accession No. ML16057A838; documents related to this amendment are 
listed in the Safety Evaluation (SE) enclosed with the amendment.
    Facility Operating License No. NPF-63: The amendment revised the 
Emergency Action Level Technical Bases document.
    Date of initial notice in Federal Register: July 21, 2015 (80 FR 
43128). The supplemental letters dated November 19, 2015, and January 
28, 2016, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in an SE dated April 13, 2016.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket Nos. 50-003, 50-247, and 50-
286, Indian Point Nuclear Generating Unit Nos. 1, 2, and 3, Westchester 
County, New York

    Date of amendment request: June 16, 2015.
    Brief description of amendments: The amendments revised the Cyber 
Security Plan Milestone 8 full implementation date by extending the 
full implementation date from June 30, 2016, to December 31, 2017.
    Date of issuance: April 12, 2016.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days of issuance.
    Amendment Nos.: 59 (Unit No. 1), 284 (Unit No. 2), and 260 (Unit 
No. 3). A publicly-available version is in ADAMS under Accession No. 
ML16064A215; documents related to these amendments are listed in the 
Safety Evaluation enclosed with the amendments.
    Provisional Operating License No. DPR-5 and Facility Operating 
License Nos. DPR-26 and DPR-64: The amendments revised the Provisional 
Operating License for Unit No. 1 and the Facility Operating Licenses 
for Unit Nos. 2 and 3.
    Date of initial notice in Federal Register: August 4, 2015 (80 FR 
46348).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 12, 2016.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-317 and 50-318, Calvert 
Cliffs Nuclear Power Plant, Units 1 and 2, Calvert County, Maryland

    Date of amendment request: November 5, 2015.
    Brief description of amendments: The amendments revise the 
Surveillance Requirement (SR) frequencies for SRs 3.4.6.4, 3.4.7.4, 
3.4.8.3, 3.5.2.10, 3.6.6.9, 3.9.4.2, and 3.9.5.4. The changes to the SR 
frequencies relocate the frequencies to the Surveillance Frequency 
Control Program.
    Date of issuance: April 11, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 317 and 295. A publicly-available version is in 
ADAMS under Accession No. ML16060A401; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Renewed Facility Operating License Nos. DPR-53 and DPR-69: 
Amendments revised the Renewed Facility Operating Licenses and 
Technical Specifications.
    Date of initial notice in Federal Register: January 5, 2016 (81 FR 
261).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 11, 2016.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: March 23, 2015, as supplemented by 
letters dated January 8, 2016, and March 21, 2016.
    Brief description of amendment: The amendment revised the technical 
specifications (TS) and relocated the secondary containment bypass 
leakage paths table from the TS to the Technical Requirements Manual.
    Date of issuance: April 19, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment No.: 156. A publicly-available version is in ADAMS under 
Accession No. ML16088A053; documents related to this amendment is 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-69: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: September 29, 2015 (80 
FR 58517). The supplemental letters dated January 8, 2016, and March 
21, 2016, provided additional information that clarified the 
application, did not expand the scope of the application as originally 
noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 19, 2016.
    No significant hazards consideration comments received: No.

Florida Power & Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No. 2 (PSL-2), St. Lucie County, Florida

    Date of amendment request: December 30, 2014, as supplemented by 
letters dated March 23, June 2, June 18, July 30, October 2, November 
3, 2015; and December 8, 2015.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to allow the use of AREVA fuel and AREVA 
M5[supreg] material as an approved fuel rod cladding at PSL-2.
    Date of issuance: April 19, 2016.
    Effective date: As of the date of issuance and shall be implemented 
upon the start of the PSL-2 Cycle 23 spring 2017 refueling outage to 
support the AREVA fuel transition project plan.
    Amendment No.: 182. A publicly-available version is in ADAMS under 
Accession No. ML16063A121; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-16: Amendment revised 
the Renewed Facility Operating License and TSs.
    Date of initial notice in Federal Register: June 9, 2015 (80 FR 
32620). The supplements dated June 2, June 18, July 30, October 2, 
November 3, and December 8, 2015, provided additional information that 
clarified the application, did not expand the scope of the application 
as originally noticed, and did not change the staff's original proposed 
no significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 19, 2016.
    No significant hazards consideration comments received: No.

[[Page 28904]]

Pacific Gas and Electric Company (PG&E), Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo 
County, California

    Date of application for amendments: June 26, 2013, as supplemented 
by letters dated September 29, October 27, October 29, November 26, and 
December 31, 2014; February 25 (two letters), May 7, October 15, and 
December 31, 2015; and January 28, 2016.
    Brief description of amendments: The amendments permit the PG&E 
(the licensee) to adopt a new fire protection licensing basis based on 
National Fire Protection Association (NFPA) Standard 805, 
``Performance-Based Standard for Fire Protection for Light Water 
Reactor Generating Plants (2001 Edition),'' at Diablo Canyon Power 
Plant, Units 1 and 2, that complies with the requirements of 10 CFR 
50.48(a) and (c) and the guidance in Revision 1 of Regulatory Guide 
1.205, ``Risk Informed Performance-Based Fire Protection for Existing 
Light-Water Nuclear Power Plants,'' December 2009.
    Date of issuance: April 14, 2016.
    Effective date: As of its date of issuance and shall be implemented 
as described in the transition license conditions.
    Amendment Nos.: Unit 1--225; Unit 2--227. A publicly-available 
version is in ADAMS under Accession No. ML16035A441; documents related 
to these amendments are listed in the Safety Evaluation enclosed with 
the amendments.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: December 26, 2013 (78 
FR 78408). The supplemental letters dated October 3, 2013; September 
29, October 27, October 29, November 26, and December 31, 2014; 
February 25 (two letters), May 7, October 15, and December 31, 2015; 
and January 28, 2016, provided additional information that clarified 
the application, did not expand the scope of the application as 
originally noticed, and did not change the staff's original proposed no 
significant hazards consideration determination as published in the 
Federal Register.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 14, 2016.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Docket Nos. 52-025 and 52-026, 
Vogtle Electric Generating Plant (VEGP), Units 3 and 4, Burke County, 
Georgia

    Date of amendment request: September 1, 2015.
    Brief description of amendment: The amendment authorized changes to 
the VEGP Units 3 and 4 plant specific emergency planning inspections, 
tests, analyses, and acceptance criteria (ITAAC) in Appendix C of VEGP 
Units 3 and 4 Combined Operating Licenses (COLs). The changes authorize 
the removal of the copy of Updated Final Safety Analysis Report Table 
7.5-1, ``Post-Accident Monitoring System'' from ITAAC in Appendix C of 
the VEGP Units 3 and 4 COLs.
    Date of issuance: March 30, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days of issuance.
    Amendment No.: 47. A publicly-available version is in ADAMS under 
Accession No. ML16061A220; documents related to these amendments are 
listed in the Safety Evaluation enclosed with the amendments.
    Facility Combined Licenses Nos. NPF-91 and NPF-92: Amendment 
revised the Facility Combined Licenses.
    Date of initial notice in Federal Register: October 27, 2015 (80 FR 
65807).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 30, 2015.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Unit Nos. 1 and 2, Appling County, Georgia

    Date of amendment request: January 13, 2015, as supplemented by 
letters dated June 16 and November 24, 2015.
    Brief description of amendments: The amendments adopt Technical 
Specification Task Force change number 523, Revision 2, ``Generic 
Letter 2008-01, Managing Gas Accumulation,'' for the Hatch Nuclear 
Plant, Unit Nos 1 and 2, technical specifications. The change revised 
or added surveillance requirements to verify that the system locations 
susceptible to gas accumulation are sufficiently filled with water and 
to provide allowances which permit performance of the verification.
    Date of issuance: April 14, 2016.
    Effective date: As of the date of issuance and shall be implemented 
within 120 days of issuance.
    Amendment Nos.: 278 and 222. A publicly-available version is in 
ADAMS under Accession No. ML16090A174; documents related to these 
amendments are listed in the Safety Evaluation enclosed with the 
amendments.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Facility Operating Licenses and Technical Specifications.
    Date of initial notice in Federal Register: March 17, 2015 (80 FR 
13911). The supplemental letters dated June 16 and November 24, 2015, 
provided additional information that clarified the application, did not 
expand the scope of the application as originally noticed, and did not 
change the staff's original proposed no significant hazards 
consideration determination as published in the Federal Register.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 14, 2016.
    No significant hazards consideration comments received: No.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: September 23, 2015.
    Brief description of amendment: The amendment revised the diesel 
generator (DG) full load rejection test and endurance and margin test 
specified by Technical Specification (TS) 3.8.1, ``AC [Alternating 
Current] Sources--Operating,'' Surveillance Requirements (SR) 3.8.1.10 
and 3.8.1.14, respectively. The change adds a new Note to SR 3.8.1.10 
and SR 3.8.1.14, consistent with Technical Specification Task Force 
(TSTF) traveler TSTF-276-A, Revision 2, ``Revise DG full load rejection 
test.'' The Note allows the full load rejection test and endurance and 
margin test to be performed at the specified power factor with 
clarifications addressing situations when the power factor cannot be 
achieved.
    Date of issuance: April 15, 2016.
    Effective date: As of its date of issuance and shall be implemented 
within 90 days of issuance.
    Amendment No.: 215. A publicly-available version is in ADAMS under 
Accession No. ML16081A194; documents related to this amendment are 
listed in the Safety Evaluation enclosed with the amendment.
    Renewed Facility Operating License No. NPF-42. The amendment 
revised the Operating License and Technical Specifications.
    Date of initial notice in Federal Register: November 24, 2015 (80 
FR 73242).

[[Page 28905]]

    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 15, 2016.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 2nd day of May 2016.

    For the Nuclear Regulatory Commission.
Anne T. Boland,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2016-10949 Filed 5-9-16; 8:45 am]
 BILLING CODE 7590-01-P